Method for immobilizing mixed waste chloride salts containing radionuclides and other hazardous wastes

ABSTRACT

The invention is a method for the encapsulation of soluble radioactive waste chloride salts containing radionuclides such as strontium, cesium and hazardous wastes such as barium so that they may be permanently stored without future threat to the environment. The process consists of contacting the salts containing the radionuclides and hazardous wastes with certain zeolites which have been found to ion exchange with the radionuclides and to occlude the chloride salts so that the resulting product is leach resistant.

CONTRACTUAL ORIGIN OF THE INVENTION

The U.S. Government has rights in this invention pursuant to ContractNo. W-31-109-ENG-38 between the U.S. Department of Energy and ArgonneNational Laboratory.

BACKGROUND OF THE INVENTION

This invention relates to a method for immobilizing radioactive wastesfor permanent disposal. More particularly, the invention relates to amethod of immobilizing mixed waste chloride salts containingradionuclides and other hazardous materials for permanent disposal.

The recovery of fissionable materials such as uranium and plutonium fromspent nuclear reactor fuels can be carried out by electrorefining methodusing electrochemical cells of the type described in U.S. Pat. Nos.4,596,647 and 2,951,793, as well as U.S. Pat. No. 4,880,506. It is theelectrorefining method which is being developed for the reprocessing ofintegral fast reactor (IFR) fuel. In a typical electrorefining cell, anelectrolyte consisting of a molten eutectic salt mixture such as KCl andLiCl is used to transport the metal or metals to be purified betweenelectrode solutions. When used to reprocess spent nuclear reactor fuels,the salt mixture becomes contaminated with radionuclides, such ascesium⁻¹³⁷ and strontium⁻⁹⁰, hazardous metals such as barium and otherspecies such as sodium and iodine⁻¹²⁹ and eventually is no longersuitable for use in the electrorefining cell.

Ideally the salt would be decontaminated by removing the heat producingradionuclides, primarily cesium and strontium, and any other metals,e.g. sodium, which could potentially interfere in the operation of theelectrorefiner and the purified salt would be recycled back to theelectrorefiner. However, the separation of cesium and strontium chloridefrom the salt is difficult, and since they are large heat producers itwould be necessary to dilute them in another matrix material and/or coolthem before they could be stored. It is therefore more practical todispose of the cesium and strontium and any other radionuclides andtoxic metal chlorides and iodides along with a portion of the saltmatrix. The waste salt containing the cesium and strontium is a highlevel waste (HLW), and as such must be disposed of in the geologicrepository for HLW. This requires that the waste form be leach resistantto prevent an uncontrolled release of the radionuclides and otherhazardous chemicals such as barium into the groundwater. Since wastesalts are chlorides and are very water soluble, a method forencapsulating and immobilizing the waste salt must be identified.

One problem with developing a waste storage medium is that the wastesalt consists primarily of chloride salts of the alkali metals and assuch is not readily amenable to treatment using procedures andtechniques developed for immobilizing the cesium and strontium in othernuclear waste streams. For instance, the chloride salts cannot be addeddirectly to glass-forming compounds and processed to yield aleach-resistant glass since glasses containing halides ions arerelatively water soluble. Therefore, for immobilization in a glassmatrix the waste chloride salts must be converted into oxides or otherchemical forms compatible with the glass-making process. However,conversion processes are expensive and time-consuming and raiseenvironmental concerns about the off-gases produced by the processes. Amortar matrix has also been considered as a possible waste form for thewaste chloride salt. A special mortar was developed to incorporatelithium, potassium, cesium and strontium chloride salts into itsstructure and thereby immobilize them. However, when irradiated, thewater in the mortar was radiolyzed and large quantities of hydrogen gaswere generated.

A new matrix for immobilizing waste chloride salts is therefore needed.Zeolites which can be treated with molten salts are potential candidatesbecause of their sorption and ion exchange properties. When somezeolites are treated with molten salts, salt molecules penetrate thecavities and channels of the zeolite and are then said to be occluded.Occluded molecules provide a transfer medium for ion exchange betweenthe cations in the zeolite and those in the bulk salt. A zeolite whichhas a high selectivity for cesium, strontium and barium would be apromising candidate for an immobilization matrix.

The ion exchange and sorption properties of zeolites in molten salts hasbeen investigated in several studies. Most of the studies, thoughinvolved nitrate salts, not chloride salts.

The ion exchange properties of several zeolites have been investigatedin molten nitrate salt solutions [C. M. Callahan, J. Inorg. Nucl. Chem.,28, 2743 (1966)]. Callahan reported the distribution coefficients(concentration in the zeolite phase/concentration in the salt phase) forsodium, calcium, potassium, rubidium, cesium and barium betweenchabazite and three solvents salts, LiNO₃, NaNO₃ and KNO₃. In eachsolvent salt one cation was preferentially sorbed, i.e., thedistribution coefficient for the preferred cation could be as much as100 times greater than that for the other cations. This study alsoshowed that ion exchange was minimal when KNO₃ was the solvent salt. Thedistribution coefficients of all the solute cations in KNO₃ were verysmall, varying from <1 to 8.7, far less than those in LiNO₃ and NaNO₃where distribution coefficients could exceed 100.

Ion exchange was investigated between salt occluded sodium A zeolite,Na₁₂ [(AlO₂)₁₂ (SiO₂)₁₂ ].10NaNO₃, and solutions of silver, lithium,potassium, cesium, thallium, calcium, and strontium in molten NaNO₃. Itwas found that silver and `presumably` lithium could be completelyexchanged with all 22 sodium ions but that exchange with the larger ionswas limited.

The occlusion of lithium, sodium and potassium nitrate in the respectiveforms of zeolite A was also studied, i.e., LiA, NaA and KA. The resultswere that LiNO₃ and NaNO₃ were readily occluded but KNO₃ was either notoccluded at all or occluded to a very limited extent.

Another study by Susic et al [J. Inorg. Nucl. Chem., 33, 2667 (1971)]investigated the salt occlusion and ion exchange properties of zeolite Ain molten alkali halides, sulfates and nitrates. This study wasprimarily concerned with measuring the amount of occluded salt for thevarious melts. It was reported that 3.5 meq of chloride per gram wasoccluded when LiCl-KCl eutectic salt was equilibrated with the lithiumform of zeolite A (LiA). The report indicated that ion exchange betweenlithium and potassium did not occur and inferred that LiCl, but not KCl,was occluded by LiA.

None of the studies cited above were directly applicable to the problemof predicting whether any zeolite would sorb cesium, strontium andbarium from a complex salt mixture such as IFR waste salt (oneconsisting primarily of LiCl and KCl with smaller amounts of NaCl) andretain them so that the zeolite with the occluded salt would act as animmobilization matrix for the cesium, strontium and barium as well asfor the matrix (lithium, potassium and sodium) chloride salts. Ingeneral the prior art teaches that steric factors are very important inion exchange and salt occlusion in molten salt-zeolite systems. Thepresence of an excess quantity of a large ion such as potassium saltwill cause the exclusion of other large ions such as cesium and barium.Occlusion and hence the desired ion exchange with cesium, strontium andbarium will not occur.

BRIEF SUMMARY OF THE INVENTION

We have discovered that, contrary to the teaching of the prior art,treating certain zeolites with molten IFR salt can provide a simple andeffective method for encapsulating waste chloride salt containingradionuclides and other hazardous materials such as barium within thealuminosilicate matrix, thereby effectively immobilizing the waste saltso that it can be permanently stored without endangering theenvironment.

The method of the invention for decontaminating and immobilizing a mixedmolten waste chloride salt containing radionuclides and other hazardousmaterials for permanent disposal comprises contacting the molten saltcontaining the radionuclides and hazardous material with a zeolite inthe sodium or lithium form, said zeolite containing cavities and beingselected from the group consisting of zeolite A, mixtures of chabaziteand erionite zeolites, and mixtures thereof, maintaining the contact fora period of time sufficient for molten salt to penetrate the cavities ofthe zeolite, thereby occluding the salt and for the radionuclides andhazardous material in the non-occluded salt to ion exchange with thesodium or lithium in the zeolite or with the occluded ions therebydecontaminating the non-occluded salt, and allowing the zeolitecontaining the radionuclides and the occluded salt to cool therebydecontaminating the non-occluded salt and immobilizing the waste saltcontaining radionuclides and hazardous materials.

Preferably any salt adhering to the zeolite is removed before thezeolite is sent to storage. This can be accomplished before the moltensalt has cooled by blowing a gas through the zeolite and salt mixture orotherwise forcibly removing the salt from the zeolite. Alternatively,the salt can be removed after cooling by washing the salt from thezeolite with water or other appropriate solvent.

The invention has the advantage over prior art methods for immobilizingwaste salts containing radionuclides and other hazardous materials inthat no complicated, costly separation steps are necessary todecontaminate the waste salt. The result of the molten salt-zeoliteequilibration is a leach resistant aluminosilicate matrix for thebarium, cesium and strontium wherein these ions are present as eitheraluminosilicates or as occluded salt molecules. Any other alkali metalions or chloride ions sorbed by the zeolite are also immobilized. Thus,by the method of the invention, normally very soluble chloride salts areconverted to an insoluble form.

It is therefore one object of the invention to provide a method ofimmobilizing waste salts containing radionuclides and other hazardousmaterials.

It is another object of the invention to provide a process forencapsulating waste chloride salts containing radionuclides and otherhazardous materials in order to immobilize the radionuclides, thehazardous materials and the soluble chloride salts so that they may bepermanently stored without posing a threat to the environment.

Finally it is the object of the invention to provide a method ofencapsulating and immobilizing soluble waste chloride salts containingcesium, strontium and barium so that they may be permanently storedwithout endangering the environment.

DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT

These and other objects of the invention for encapsulating andimmobilizing radioactive cesium, strontium and barium ions contained ina molten waste chloride salt mixture for permanent disposal may be metby contacting the molten waste chloride salt containing the cesium,strontium, barium and other waste salt components with dehydratedzeolite A in the sodium form, said zeolite containing cavities,maintaining the contact for a period of time sufficient for the salt topenetrate the cavities in the zeolite, thereby occluding the salt withinthe zeolite and for the cesium, strontium, and barium in thenon-occluded salt to ion-exchange with the sodium in the zeolite or withcations in the occluded salt, removing the non-occluded salt from thezeolite, and cooling the zeolite containing the cesium, strontium andbarium and the occluded salt thereby immobilizing the waste salt forpermanent storage.

Contact between the zeolite and the contaminated molten salt may takeplace by passing the molten salt through a packed column of thedehydrated zeolite, maintaining contact until the cesium, strontium andbarium are sorbed by ion-exchanging with the cations in the zeolite orthe occluded salt, and collecting the decontaminated salt for recyclingback to the process. Alternatively, sufficient molten salt may be addedto the dehydrated zeolite to form an occluded salt-zeolite compoundwhich sorbs the cesium, strontium and barium contained in the salt in abatch type process.

The preferred zeolite for use with this process is zeolite A which hasbeen found to be highly selective for cesium, strontium and barium in achloride salt solution and which also sorbs lithium and potassium to arelatively lesser extent. Zeolite A is also preferred because of itsability to occlude large quantities of chloride salts, even in thepresence of high concentrations of potassium in the molten salt. Anothersuitable zeolite is zeolite Ionsiv™ IE95 which is a mixture ofchabazite- and erionite-type zeolites. This zeolite has good selectivityfor cesium and occludes less salt than zeolite A. Because of its lowerselectivity for barium and strontium, it may be preferable to use amixture of zeolite IE95 and zeolite A. Such a mixture may be used whenit is desired to minimize the amount of occluded salt. Both zeolitesform very leach resistant salt occluded zeolite compounds when they sorbsalt molecules.

Actual IFR waste salt will contain, in addition to the cesium andstrontium, comparatively small amounts of other radionuclides such asthe rare earths and iodide. It may be possible to use zeolites A or IE95to immobilize them, but other zeolites may prove more suitable becauseof steric and charge density factors. Iodide ion is larger than cesiumand the rare earths cations are normally trivalent.

The temperature of the molten salt must be sufficient to maintain thesalt in a liquid state so ion exchange of the radionuclides can takeplace and so that the salt can penetrate into the cavities of thezeolite. Generally a temperature of about 375° C. will providesufficient liquidity for the salt to flow into the zeolite. Preferably,temperatures over about 600° C. are to be avoided since they may resultin destruction of the zeolite.

Preferably the zeolite is dehydrated before contact is made with thesalt. This is done to minimize the amount of water in the final product,since water may radiolyze during storage of the radioactive mediaproducing hydrogen gas. The zeolites can be easily dehydrated with aheated inert gas purge or vacuum degassing.

The following examples are given to illustrate the invention and are notto be taken as limiting the scope of the invention which is defined bythe appended claims.

EXAMPLE I

The procedures given below were used for the following Examples. Theexperiments were run at 400°±25° C. in an argon atmosphere glove box.The composition of the simulated IFR waste salt was 92.5 wt % LiCl-KCleutectic salt, 5 wt % NaCl, 1 wt % CsCl, 1 wt % BaCl₂ and 0.5 wt %SrCl₂. The zeolites were dehydrated by heating the zeolite to 350° C.with a gaseous purge of dry nitrogen.

The ion exchange properties of several zeolites, A, IE95, mordenite,clinoptilolite, and X were studied as follows. About 12 g of each of thevarious zeolites, having a nominal particle size of about 2 microns,were packed into a column and contacted with about 35 g of simulated IFRwaste salt at 400°±25° C. The molten salt was allowed to filter througheach column under gravity for 2-4 days. Salt which flowed through theentire length of the column was collected. After several grams of saltwere collected, the experiments were terminated. The column and the saltin the collection crucible were quickly cooled to room temperature. Thezeolite was washed to remove adhering surface salt, dried and thenanalyzed. Inductively coupled plasma spectroscopy (ICP) was used toanalyze for barium, lithium, potassium, silicon, sodium and strontium.Atomic emission spectroscopy was used to analyze for cesium. Thecompositions of the filtered salt samples from the various runs and thecomposition of the starting salt are given in Table 1 below.

                  TABLE 1                                                         ______________________________________                                        Composition of Salt Filtered through Various                                  Zeolite Columns and Simulated IFR Waste Salt                                  Zeolite Concentration in Filtered Salt (wt %)                                 in Column                                                                             Ba     Cs       K     Li    Na    Sr                                  ______________________________________                                        A       0.001  <0.001   26.5  5.00  6.36  <0.001                              IE95    0.945   0.053   18.6  6.25  4.80  0.385                               Mordenite                                                                             1.00   0.07     20.2  6.25  6.49  0.469                               Clinop- 0.659  0.53     18.8  6.19  5.25  0.334                               tilolite                                                                      Starting                                                                              0.64   0.75     24.0  7.15  1.80  0.26                                Salt Com-                                                                     position                                                                      ______________________________________                                    

The data demonstrate that zeolite A preferentially sorbs barium, cesiumand strontium since the salt which passed through the zeolite A columncontained essentially none of these ions. However, the salt which flowedthrough the IE95, mordenite and clinoptilolite columns, contained morebarium and strontium than the starting salt and less potassium andcesium than the starting salt. This indicates that these zeolites had alower selectivity for the divalent ions than for potassium and cesium.The sodium concentration is higher in all the filtered salt samples thanin the starting salt. This indicated that the sodium in the variouszeolites has been exchanged with cations in the salt.

EXAMPLE II

The various zeolites from Example I were analyzed to determine theirbarium, cesium and strontium content. The analytical results forsamples, taken from the middle of the zeolite column, are given in Table2 below. Also included in Table 2 are the concentrations of lithium,potassium and sodium.

                  TABLE 2                                                         ______________________________________                                        Composition of Several Zeolites Treated                                       with Simulated IFR Waste Salt                                                          Concentration (wt %)                                                 Zeolite  Ba      Cs      K     Li    Na    Sr                                 ______________________________________                                        A        1.40    2.0     7.42  4.88  0.73  0.58                                        1.41    1.8     7.81  5.30  0.80  0.37                                        1.38    1.8     7.79  5.25  0.79  0.36                               IE95.sup.a                                                                             0.72    1.9     9.84  0.49  0.41  0.28                               Mordenite                                                                              0.38    1.7     7.61  0.02  0.17  0.09                               Clinop-  0.18    1.1     8.27  1.21  0.51  0.18                               tilolite                                                                      X        1.32    0.2     6.73  1.77  2.83  0.62                               ______________________________________                                         .sup.a A mixture of chabazite and erionite marketed by Union Carbide as       Ionsiv ™ IE95.                                                        

Note that the sum of the barium, cesium and strontium concentrations inthe salt occluded zeolite A is higher than that of any of the otherzeolites. Note also that zeolite IE95 is the only other zeolite able tosorb significant concentrations of the ions of interest.

EXAMPLE III

The salt occlusion properties of the zeolites used in the precedingExamples were measured by a determination of their chloride ionconcentration using pyrohydrolysis and ion chromatography. These dateare given in Table 3 below.

                  TABLE 3                                                         ______________________________________                                        CONCENTRATION OF CHLORIDE IONS IN                                             TREATED WITH SIMULATED IFR WASTE SALT                                         zeolite        Cl.sup.-  (wt %)                                               ______________________________________                                        A              16.9-19.2                                                      IE95           3.9                                                            Mordenite      0.8                                                            Clinoptilolite 6.6                                                            X              1.3                                                            ______________________________________                                    

As can be seen, zeolite A contained the largest concentration ofchloride ion after treatment with the molten salt than any of the otherzeolites. The 16.9 to 19.2 wt % measured is equivalent to 4.8 to 5.3 meqchloride ion per gram of salt occluded zeolite. IE95 and clinoptiloliteoccluded moderate amount of chloride ion but mordenite and zeolite Xoccluded practically none.

EXAMPLE IV

12 g of dehydrated zeolite A in the sodium form (NaA) with a nominalparticle size of 2 microns was contacted with about 35 g of simulatedIFR waste salt at 400°±25° C. as described in Example 1. The zeolite waswashed, dried and analyzed. The results of the analyses for a singlesample taken from the top 1/2 inch of the column are reported in Table 4below. The concentration of barium, cesium and strontium are higher atthe top of the column than in the middle as in Example II. This isfurther evidence of an ion exchange process. The relative selectivity ofzeolite A for the components of the salt can be seen by comparing theconcentration of the cations (in mole %) in the salt occluded zeoliteand in the simulated IFR waste salt as shown in Table 4. Also includedin the table is the composition of anhydrous zeolite NaA.

                  TABLE 4                                                         ______________________________________                                        Concentration of Metal Ions                                                            Concen-                                                                       tration   Metal Ions                                                 Sample   Units     Ba     Cs   K    Li   Na   Sr                              ______________________________________                                        Salt     wt %      4.83   2.7  6.58 4.30 0.49 3.18                            Occluded                                                                      Zeolite  meq/g.sup.a                                                                             0.70   0.20 1.70 6.20 0.21 0.73                                     mole %.sup.b                                                                            3.9    2.2  18.7 68.9 2.3  4.0                             Anhydrous                                                                              wt %                            16.2                                 NaA                                                                                    meq/g.sup.c                     7.00                                 Simulated                                                                              wt %      0.64   0.75 24.0 7.15 1.80 0.26                            IFR Waste                                                                     Salt     mole %    0.27   0.32 35.3 59.5 4.5  0.17                            ______________________________________                                         .sup.a g refers to gram of salt occluded zeolite. Calculated by dividing      the amount of each cation in one gram of salt occluded zeolite by the         equivalent weight.                                                            .sup.b Calculated by dividing the number of mmoles/g for each cation by       the total number of mmoles of cations, 9.03; zeolite component is             excluded.                                                                     .sup.c g refers to gram of anhydrous zeolite.                            

These data show that NaA functions as an effective ion exchanger forcesium, strontium and barium in molten simulated IFR waste salt sorbingfrom 2-4 mol % of each. Note also that most of the sodium initiallypresent in the NaA was replaced in the salt occluded zeolite by theother metal ions in the molten salt.

EXAMPLE V

The salt which filtered through the column of zeolite A in Example IVwas analyzed. Decontamination factors (concentration in/concentrationout) were calculated from this analysis and the analysis of the startingsalt. These are given in Table 5 below.

                  TABLE 5                                                         ______________________________________                                        Concentration of Metal Ions in Filtered and                                   Simulated IFR Waste Salt (wt %)                                                        Ba    Cs      K      Li   Na    Sr                                   ______________________________________                                        Filtered   0.001   <0.001  26.5 5.00 6.36  <0.001                             Salt                                                                          IFR Salt   0.64    0.75    24.0 7.15 1.80  0.26                               Decontamination                                                                          640     >750    0.9  1.4  0.3   >260                               Factor                                                                        ______________________________________                                    

The decontamination factors for barium, cesium and strontium are about640, >750 and >260 respectively. The decontamination factors forpotassium and lithium are 0.9 and 1.5 respectively, i.e., theconcentration of potassium is slightly higher in the filtered salt thanin the simulated IFR waste salt while the concentration of lithium islower.

EXAMPLE VI

Partial evidence for the occlusion of salt molecules was obtained from adetermination of the chloride ion concentration of the salt occludedzeolites as in Example III. Confirmatory evidence for salt occlusion inzeolite A was obtained from an analysis of the cation concentration ofsalt occluded zeolite A. The total concentration of cations was obtainedby summing the number of milliequivalents (meq) for all the metal ions,or 9.74 meq/g of salt occluded zeolite as in Example IV. The siliconconcentrations in anhydrous NaA (19.8 wt %) and in the salt occludedzeolite (13 wt %) were used to measure the amount of anhydrous zeolitepresent in the salt occluded zeolite compound, 13/19.8 or about 66 wt %.In 0.66 g of anhydrous zeolite NaA, there were 4.65 meq of sodium. Thenumber of excess cations was 9.75-4.65 or 5.10 meq/g. Thus, both thecation and chloride analyses showed that 5.0±0.3 meq of salt wasoccluded per gram of the salt occluded zeolite compound.

                  TABLE 6                                                         ______________________________________                                        Concentrations of Metal Ions and Silicon                                      wt % (meq/g)                                                                  Sample                                                                              Ba      Cs      K     Li    Na    Sr    Si                              ______________________________________                                        Salt  1.45    2.0     7.36  4.89  0.73  0.661 13.0                            Occ-  (0.22)  (0.15)  (1.90)                                                                              (7.00)                                                                              (0.32)                                                                              (0.16)                                luded                                                                         Zeolite                                                                       ______________________________________                                    

EXAMPLE VII

Samples of the washed and dried occluded salt-zeolite compound wereleached in deionized water for periods of time ranging from 1 day to 42days. After the leach period, the mixture was filtered with a 0.2μfilter. The leachate was acidified and analyzed. The zeolite was driedand analyzed. The details of the analyses for two samples consisting ofa 50-50 wt % mixture of zeolite A and IE95 are given in Table 7. Theleach periods for these two experiments were 1 day and 42 days.

                  TABLE 7                                                         ______________________________________                                        Leach Test Results                                                                       Duration                                                           Sample Zeolite (g)                                                                             (d)      Ba     Cs     Sr                                    ______________________________________                                                        Concentration (wt %)                                          1      0.216     1        2.00   1.8    0.92                                  2      0.233     42       1.65   1.8    0.71                                  Leachate (ml)     Concentration (μg/ml)                                    1      21.65     1        <0.02  0.1    0.01                                  2      23.36     42       <0.02  0.07   0.015                                                 Amount Released (%)                                           1                         <0.01  0.06   0.01                                  2                         <0.01  0.04   0.02                                  ______________________________________                                    

The percentages of barium, cesium and strontium released in the one daytest were <0.01, 0.06 and 0.01, respectively and in the 42 day test,they were <0.01, 0.04 and 0.02, respectively. These release rates aresignificantly smaller than the 2.3% reported in a previous study. [M.Liquornik and Y. Marcus, Israel J. Chem 6, 115 (1968)]. Thus, bothzeolites A and IE95 equilibrated with molten simulated IFR waste saltact as a particularly good leach resistant or immobilization matrix forthe barium, cesium and strontium. Other experiments with zeolite A aloneyielded similar results.

As can be seen from the preceding discussion and Examples, the method ofthe invention provides a safe and effective manner in which toimmobilize highly soluble waste chloride salts containing radioactivenuclides to enable the salts to be sent to permanent storage withoutfear of any detrimental effect on the environment.

The embodiment of the invention in which an exclusive property orprivilege is claimed is defined as follows:
 1. A method ofdecontaminating and immobilizing molten waste chloride salts containingradionuclides and barium for permanent disposal comprising:contactingthe molten salt containing the radionuclides and barium with a zeolitein a form selected from the group consisting of sodium and lithium, saidzeolite containing cavities and being selected from the group consistingof zeolite A, mixtures of chabazite and erionite zeolites, and mixturesthereof, maintaining the contact for a period of time sufficient formolten salt to penetrate the cavities of the zeolite, thereby occludingthe salt and for the radionuclides and barium in the non-occluded saltto ion exchange with the sodium or lithium in the zeolite therebydecontaminating the non-occluded salt, and allowing the zeolitecontaining the radionuclides and the occluded salt to cool therebydecontaminating the non-occluded waste salt and immobilizing the wastesalt containing radionuclides and other hazardous material.
 2. Themethod of claim 1 wherein the zeolite is zeolite A.
 3. The method ofclaim 2 wherein the radionuclides are one or more members selected fromthe group consisting of cesium, and strontium.
 4. The method of claim 3wherein the chloride salt consists of lithium chloride, potassiumchloride.
 5. The method of claim 4 wherein the chloride salt alsocontains sodium chloride.
 6. The method of claim 5 wherein any excesssalt is removed from the zeolite before the salt is cooled.
 7. Themethod of claim 5 wherein any excess salt is removed from the zeoliteafter the salt is cooled by washing the zeolite containing the exchangedradionuclides and the occluded salt with water.
 8. The method of claim 5wherein the temperature of the molten salt is between 375° and 600° C.9. A method of immobilizing and decontaminating a waste chloride saltmixture of lithium and potassium containing sodium, cesium, strontium,and barium in addition to other radionuclides and hazardous materialscomprising:contacting the molten waste salt mixture with dehydratedzeolite A in the sodium form, said zeolite containing cavities,maintaining the contact for a period of time sufficient for molten saltto penetrate the cavities of the zeolite, thereby occluding the salt andfor the cesium, strontium and barium in the non-occluded salt to ionexchange with the sodium in the zeolite, thereby decontaminating thenon-occluded salt, removing the unoccluded salt from the zeolite, andcooling the zeolite containing the cesium, strontium and barium and theoccluded salts thereby decontaminating the non-occluded salt andimmobilizing the chloride salts containing cesium, strontium and barium.